ML20140C742
| ML20140C742 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 06/11/1984 |
| From: | Musolf D NORTHERN STATES POWER CO. |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| TAC-53456, TAC-53457, NUDOCS 8406190426 | |
| Download: ML20140C742 (7) | |
Text
1 Northem States Power Company 414 Nicollet Mall Minneapohs. Minnesota 55401 Telephone (612) 330-5500 June 11, 1984 Director Office of Nuclear Reactor Regulation U S Nuclear Regulatory Commission
,Uashington, DC 20555 PRAIRIE ISLAND NUCLEAR GENERATING PLANT Docket Nos. 50-282 License Nos. DPR-42 50-306 DPR-60 Response to NRC Request for Supplemental Information on the Second Ten Year ISI Program Request for Relief Attached, as a supplement to our March 9,1984 letter, is our response to the NRC Staff's request for supplemental information on the Second Ten Year ISI Program Requests for Relief submitted to the Commission by our letter dated December 22, i983.
Please contact us if you have any questions related to this supplemental info rmation.
DS w.
David Musolf Manager - Nuclear Sup t Services DMM/EFE/bd c: J G Keppler (w/ attachment)
NRR Project Manager, NRC (w/ attachment)
Resident Inspector, NRC (w/ attachment)
G Charnoff (w/ attachment)
Attachment k
-f ' \\
t 8406190426 040611 PDR ADOCK 05000282 O
)
.Directtr of NRR Jun2 11, 1984 Attachment Response to NRC Request for Supplemental Information Item #1:
No additional references or documents should be needed to aid your review.
Item #2:
Request for Relief No. 45 requests that volumetric examina-tion of welds on the Regenerative Heat Exchangers Excess Let-down Heat Exchangers, and RHR Heat Exchangers be examined in at.cordance with the procedure for pipe welds rather than for thick-walled vessels since the components are fabricated from piping and thin plate.
If relief is granted, will 100% of the code required volume (CRV) for thin walled components be examined for all of the welds?
Answer:
The NSP Procedure for Ultrasonic Examination of Pipe Welds utili-zes a minimum of 1 1/2 node metal path examination.
The required scanning area is defined as "the greater of 3t or 3 inches" from the toe of the weld on each side, to the extent practical pre-cluding any geometric limitations.
The code required volume (CRV) for thin-walled components as deter-mined by Section XI; "the weld + t either side", will be more than covered by the NSP Piping Procedure.
Item #3:
Request for Relief No. 48 requests permission to use a "back reflection" ultrasonic examination method similar to the method used for the baseline examination to examine the Reactor Cool-ant Pump Flange Bolting rather than the procedure specified in Article 5,Section V.
Please provide a reference for the "back reflection" examination method to be used or a copy of the writ-ten procedure if trie method is not documented in material gener.
ally available to the NRC.
Answer:
A qualification program was initiated in response to similar questions concerning this item in our original PI-ISI Submittal for the first l
ten year inspection interval.
During the qualification test it was demonstrated that the NSP-UT-4 procedure, which utilizes a back re-flection technique for flaw evaluation, was a more sensitive examina-tion than the technique specified in ASME Section V, Article 5, Para-graph T-525.2.
The results indicated that at the same nominal metal path, the NSP procedure was approximately 6db more sensitive than the ASME tech-nique (See Figure 1).
In addition to the percent-of-DAC reporting l
level, the NSP procedure dictates that any reflector, regardless of amplitude, which is accompanied by a 50% loss of back wall reflec-tion must be reported / evaluated. As poorer end reflecting surfaces are encountered, the NSP procedure tends to become a much more con-servative approach to bolt and stud examination.
l l
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. Director of NRR June 11, 1984 Attachment Figure 1 Qualification Program Results:
ASME V vs NSP Procedure 10 0 "
@ ~~
FBH @
Stud SR go 3" MP o @ 15" MP lo Go
% FSH Sa ASME Reporting (DAC) Level i
%s - -
FBH @
3a 12n gp NSF Reporting Level 20 (20% of Stud BR)
N>
j z
1 2 3 4 5 6 l 8 1 10 Screen Divisions Item #4:
Request for Relief No. 50 requests relief from examination of 67 Class 2 welds, 1 Class 1 weld, and 22 Class 2 sup-ports on Unit 1 and 90 Class 2 welds and 28 Class 2 sup-ports on Unit 2.
All of the specified examination areas are inaccessible because they are encapsulated in guard pipe or embedded in concrete.
Please state the criteria that was used for including these welds in the examination sample. Is it possible that inspection of these welds is not required or that alternate accessible welds could be selected within the ASME code guidelines such that the scope of the relief request could be reduced?
Answer:
The criteria used for including these welds in the examination sample is IWB-2500 and IWB-2600 for Class 1 and IWC-1220 Table IWC-2520 and Paragraph IWC-2411 for Class 2 as referenced by 10CFR50.55a to the 1974 Edition Summer 1975 Addenda of Section XI.
Reduction of this request for relief is possible, however for those items listed below no reduction can be made.
Code requirement: '74 Summer '75, Paragraph IWC-2411 for Class II 1.
item (b)..... equivalent to.. 100%...in one of the multiple streams....
2.
item (c)(2)..approximately equal number...in each.. stream...
3.
item (c)(3).. select different components in each... stream...
4.
item (e)..... divided among the number of inspection inter-vals...
5.
item (e)(3)..at least one examination in each stream...
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Direct r of NRR
'Jun] 11, 1984 g,
Attachment Unit 1
/
System Size No. of No. Accessible No. over Per item No. Above Welds Welds 40 years 1.
2.
3.
4.
5.
Main Steam A Loop 31" 10 10 10 Yes No No Yes No B Loop 31" 10 Main Steam Fittings A Loop 8
6 7
Yes No No Yes Yes B Loop 6
2 Feedwater A Loop 16" 22 22 22 Yes No No Yes Yes B Loop 16" 21 7
Containment Sump B No No No No A Loop 14" 4
0 0
No B Loop 14" 4
0 A Loop 14" 2
1 2
Yes Yes Yes Yes No B Loop 14" 2
1 Unit II c
System Size No. of No. Accessible No. over Per item No. Above Welds Welds 40 years 1.
2.
3.
4.
5.
Main Steam A Loop 30" 10 10 10 Yes No No Yes Yes B Loop 30" 9
4 A Loop 31" 12 10 12 No No No Yes Yes B Loop 31" 12 1
Main Steam Fittings A Loop 9
8 8
Yes No No Yes Yes B Loop 7
2 lFeedwater A' Loop 16" 19 19 20 Yes No No Yes Yes B Loop 16" 20 5
Containment Sump B A Loop 14" 3
0 4
No No No No No B Loop 14" 4
0 l
A Loop 12" 2
1 2
Yes Yes Yes Yes Yes l
x B Loop 12" 2
1 The component supports listed for Class II are embedded in concrete and are the only ones on the particular line size.
(Containment Sump A & B Discharge)
. Unit 1 & 2 Class I percentage required by IWB-2500 is not feasible.
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Directar of NRR June 11, 1984 Attachment The remaining component supports listed for Class Il are encapsulated by guard pipe and are uninspectable for surface examination as required in Table IWC-2500-1 item C3.20, however the VT-3 and VT-4 examinations re-quired by Table IWF-2500-1 can be completed.
The remaining Class 2 weld not identified above could be deleted from the request for relief by using the criteria items listed above, this would re-duce the number of welds by 20 and 49 for Units I and II respectively.
I i
Item #5:
Request for Relief No. 66 requests relief from volumetric ex-amination of the nozzle inner radiused sections for a number of key nozzles on the Steam Generators, Pressurizer, Regener-ative Heat Exchanger, Excess Letdown Heat Exchangers and Ac-cumulators due to the unavailability of suitable examination methods that have been explored in attempting to find suitable methods for examination of the nozzle inner radius sections.
Answer:
Industries' historical evidence indicates inner radius areas of cer-tain kinds of nozzles have been susceptible to thermal fatigue-type I
fractures resulting from temperature gradients in highly cyclic en-vironments.
The intent of the code requirement is to provide some minimum criteria for monitoring these inner radius sections of noz-zles which are subject to these ::onditions such that thermal degrada-tion may be detected in the early stages. However, the nozzles listed in the latest revision of Request for Relief #66 do'not experience these highly cyclic conditions. Therefore, we postulate that an en-gineering basis for the examination of these nozzles does not exist.
f In addition, there is no conclusive volumetric method currently avail-f able which has been qualified as adequate to assess the condition of the specified examination volumes; evidenced by the non-existence of i
specific technical guidelines in ASME - Sections V or XI. This is due primarily to the complex geometrical configurat' ions of the nozzles, the compound angles and lengthy metal paths of the sound beams, the limited scanning areas, component surface finishes and material types, and other ultrasonic considerations associated with an inspection of these nozzles.
Knowledgeable personnel at the EPRI-N0E Center have been consulted and are generally in agreement with these findings and conclusion.
Recent examinations of piping welds adjacent to several of the subject nozzles have resulted in relatively high man-REM exposures. The do-i sage rates at these locations have been ranging from 200 mrem /hr. to l
1000 mrem /hr. (general field) and 800 mrem /hr. to 2500 mrem /hr (con-tact). We feel, therefore, performing these examinations which are at best, questionable, would not be consistent with normal ALARA principles.
It is our intention to pursue the steam generator feedwater and the pressurizer spray nozzles' examination to the fullest extent practical, i
as these nozzles appear to be the most susceptible to a thermal fati-gue mechanism. Coincidentally, these nozzles post a minimum man-REM exposure allowing for sufficient examination time in order to address the questions concerning the adequacy of the techniques employed.
If service induced discontinuities are discovered, the remaining nozzles l
shall be assessed for possible implementation of similar techniques.
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Directsr of NRR r
' June 11, 1984 Attachment The submitted tables and Request for Relief No. 66 will be updated and formally transmitted in the next revision Enclosed is a copy of the Request for Relief No. 66 for your infor-mation, (Attachment #1).
Item #6:
Request for Relief NO. 67 requests relief from volumetric l
examination of nozzle-to-vessel welds on the Excess Letdown Heat Exchanger due to complex geometrical configuration.
Please provide sketches of the referenced welds and indicate the sources of examination methods that have been explored.
Answer:
This Request for Relief will be withdrawn based on the exemption criteria in IWB-1220. A re-evaluation of the nozzle revealed that this nozzle is a 1 inch nominal pipe size and therefore exempt i
from inspection.
.The submittal tables will be revised to reflect this change.
f 4-1 5 of 5
66.
REQUEST FOR RELIEF CODE PROGRAM CODE EXAM COMPONENT OR ITEM CLASS TABLE ITEM CATEGORY STEAM GENERATORS - PRIMARY INLET /0UTLET N0ZZLES I
3.1 B3.140 B-D MAIN STEAM II 2.10 C2.22 C-B PRESSURIZER -
RELIEF, SURGE & SAFETY N0ZZLES I
3.1 B3.120 B-D REGENERATIVE HEAT EXCHANGER N0ZZLES I
3.1 B3.160 B-D ACCUMULATOR N0ZZLES II 2.10 C2.22 C-B CODE REQUIREMENT Volumetric (ultrasonic) examinations for the nozzle inner radius sections shall be conducted in accordance with Articles IWB-2500 and IWC-2500 for Code Class I and II, respectively.
BASIS The code required volume will not be examined based on the following criteria:
A.
These nozzles do not experience high cyclic temperature gradients during nonnal operation, therefore the conditions for producing a thermal fatigue mechanism are not applicable.
B.
Presently, there is no comprehensive inspection technique available, nor guidance for such in the ASME Code, which would provide a conclusive assessment of the code required volumes of the inner radii, par-ticularly since no preservice results are available for comparison.
C.
Upon consideration of the above factors, a best effort examination approach to these nozzle inner ra-dius sections is not consistent with standard ALARA practices. An estimated 8 to 10 man-REM exposure rate over the interval, per unit, (at the present radiations levels), would be experienced in attempt-ing to perform such inconclusive examinations.
ALTERNATE The steam generator feedwater nozzles (Table 2.10) and the pressurizer spray nozzle (Table 3.1) may be suscept-ible to a thermal fatigue mechanism due to the potential for high cyclic temperature gradients, therefore an at-tempt will be made to ultrasonically examine these inner radius areas.
If service defects are detected by these examinations, the nozzles listed in the above table shall be assessed for similar examinations. Meanwhile, if a more comprehensive technique is developed and qualified, it will be implemented.
y oii SCHEDULE FOR IMPLEMENTATION g
December 16, 1983.
3 MFM10RWPXG1